1. FIELD OF THE INVENTION
This invention relates to pressure vessel penetrations and, more particularly, to feedwater inlet conduit and steam discharge nozzle apparatus for nuclear reactor systems, and the like.
2. DESCRIPTION OF THE PRIOR ART
Some designs for nuclear power reactor systems have a reactor core that is disposed within a pressure vessel. The reactor core generates heat that is removed by means of a primary coolant which flows through passageways or channels in the core. This primary coolant is then pumped from the reactor core passageways to one or more heat exchangers that are disposed within the pressure vessel. Within the heat exchangers, the heat that the primary coolant absorbed when in the reactor core is transferred to a secondary coolant.
Water is the usual secondary coolant substance, and within the heat exchangers, the water is heated until it vaporizes and forms steam. The steam from the heat exchanger flows out of the pressure vessel through a discharge nozzle to one or more steam turbines for power generation, or the like. The spent steam from the turbines then flows through a condenser which converts the vapor back into liquid water. This water--or feedwater--is reintroduced into the heat exchangers through a feedwater inlet conduit that penetrates the pressure vessel wall. Preferably, these heat exchangers have an array of generally parallel tubes that contain the primary coolant. The feedwater for the secondary coolant cycle is permitted to flow in the spaces between the tubes in order to absorb the heat from the primary coolant fluid that is flowing within these tubes.
Although the foregoing structural arrangement provides a relatively economical and efficient nuclear power system, there are a number of design difficulties. For example, the feedwater inlet conduit and the steam discharge nozzle each require separate pressure vessel penetrations. This specific need involves a great deal of expensive high-quality special machining, welding and weld joint testing. The weld joint testing, moreover, must be conducted not only during manufacture, but also at regular inspection intervals during the service life of the reactor. Further in this regard, the pressure vessel penetrations that are needed to accommodate all of these conduits and nozzles tend to weaken the pressure vessel structure and produce undesirable areas of local stress concentration.
With respect to local stress concentrations, it also should be noted that during reactor operation the relatively thick steel walls of the pressure vessel tend to reach an equilibrium temperature that approaches the temperature of the steam in the secondary coolant loop. The feedwater in the secondary coolant loop, however, that flows into the heat exchanger through the inlet conduit, is relativey cold. This temperature difference leads to stresses in the relatively thick steel walls of the pressure vessel that, if unchecked, can be extremely destructive. To overcome this problem it has been customary to insert thermal insulation between the feedwater inlet conduit and the circumscribing portion of the pressure vessel. This technique, although generally adequate, tends to increase costs by reason of the special machining, welds, fittings and inspections that are required.
Because of the expansion and contraction that metals undergo in response to temperature changes, these prior penetration designs also tend to produce troublesome differential thermal movements, the relatively hot pressure vessel, for example, expanding through a greater distance than the colder associated feedwater inlet conduit.
As hereinbefore described, some of these designs have a number of modular heat exchangers mounted within the pressure vessel. Naturally, these modules must be secured within the pressure vessel in a manner that is capable of sustaining the anticipated stresses to which each of the modules will be subjected. In view of the thermally induced expansion and contraction, shocks and stresses that could reasonably be expected during power reactor start-up, operating transients and shut-down conditions, in additon to the need to remove the modules for occasional service inspections and repair with remote handling equipment because of the radiation environment, the problem of a suitable module mounting is extremely difficult to solve. Typically, past attempts to solve this complicated support problem have involved the use of support lugs, a built-up metal ledge and bolted support plates, all within the pressure vessel. These structures, however, involve a need for costly, precise machining and tedious, carefully conducted assembly and disassembly procedures.
Consequently, the requirement to provide more effective and less expensive thermal effects protection for feedwater inlet conduits and to improve the support structure for heat exchanger modules within a reactor pressure vessel has remained unsatisified to a great extent.